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Briggs, Samuel A.

Publications and source records attributed to Briggs, Samuel A..

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Influence of processing on secondary phase formation and microstructural evolution at U-10Mo alloy and Zr interlayer interfaces

Thermo-mechanical processing of uranium-10 wt. % molybdenum (U-10Mo) fuel plates leads to microstructure changes at the U-10Mo/Zr interfaces. Secondary phases formed at this interface are particularly important to interfacial bond strength, process optimization, and maintaining structural integrity of the U-10Mo fuel plates during irradiation. In this work, we determined the phases and phase transformation products occurring at the interface of the U-10Mo fuel and Zr interlayer when the fuel plate is subjected to short and long hot isostatic pressure times. Interfacial morphology, structure and composition of phases formed, and relative hardness across the U-10Mo/Zr interfaces were studied using a multi-length scale, multi-modal characterization approach involving electron microscopy, atom probe tomography, and atomic force microscopy. Here, results highlight that the extent of phase transformations, secondary phase formation, and hardness variability across interfaces can be controlled by modifying processing parameters. Phase diagram construction and thermodynamic calculations were performed using the Thermocalc software to identify expected phases formed at interfaces during the maximum hold temperature of 560 °C experienced during HIP.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Microstructure and microchemistry changes at U-10Mo fuel/AA6061 cladding interfaces with varying hot isostatic pressing conditions

Monolithic uranium - 10 wt.% molybdenum (U-10Mo) is a promising high-assay low-enriched uranium fuel system for nuclear reactors used in research, medical isotope production, and remote power applications. During manufacturing, AA6061 cladding is bonded to the U-10Mo fuel plate via hot isostatic pressing, during which diffusion and fuel/cladding chemical interaction occurs at plate edges. Furthermore, the microstructure and microchemistry changes that occur at fuel/cladding interfaces are important to understand as a function of process parameters to develop a reliable fuel fabrication process and to meet the desired specifications. Here, microstructural and microchemical changes are studied using complementary electron microscopy and atom probe tomography for two manufactured fuel plates with varied hot isostatic pressing (HIP) parameters. Results highlight that modifying thermomechanical processing parameters significantly changed the interaction layer thickness between U-10Mo and AA6061 by an order of magnitude. In addition to this the distribution and concentration of elements (i.e., Al, Si, Mg) from cladding to fuel was also investigated.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Improved irradiation resistance of accident-tolerant high-strength FeCrAl alloys with heterogeneous structures

In this work, post–neutron irradiation examination is performed on advanced accident-tolerant fuel (ATF) cladding iron-chromium-aluminum (FeCrAl) alloys with ~10–13at. % Cr, ~10–12 at. % Al, ~1 at. % Mo, and minor alloying elements including Y irradiated to a damage level of 7 displacements per atom (dpa) at irradiation temperatures of 267–282 °C. A compositional dependency of the Cr and Al content is observed on the ratio of sessile and glissile dislocation loops, where the density of a$\langle$100$\rangle$ type loops is somewhat higher than the a/2$\langle$111$\rangle$ type loops. The α' precipitate number density is inversely correlated to the starting Cr concentration of the alloys of interest. The irradiation to a higher dose of 7 dpa results in a higher density of dislocation loops and α' precipitates for the same alloys at a lower irradiation dose, such as 1.8 dpa. In this work, the effect of α' precipitates on the dislocation loop density is discussed, and the presence of α' appears to inhibit the nucleation of loops. Compared with first-generation FeCrAl alloys, these advanced alloys with heterogeneous structure exhibit a lower Cr concentration in α' precipitation at the same dose level; they act as weaker obstacles deviating from the primary hardening contribution from the mature α'. Hence, the overall irradiation-induced hardening decreases; our alloys show improved radiation resistance because of their stronger sink strengths. The results presented in this paper could provide insights for the design and optimization of ATF cladding materials for future fission and space applications.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Deconvoluting the Effect of Chromium and Aluminum on the Radiation Response of Wrought FeCrAl Alloys After Low-Dose Neutron Irradiation

FeCrAl alloys have been extensively investigated over the past decade as a candidate material for accident-tolerant fuel cladding in light water reactors. Here, we have completed a first of its kind study where Al and Cr concentrations are varied systematically under neutron irradiation to elucidate the post irradiation microstructural and mechanical response as a direct result of alloying content. Neutron irradiations were performed on alloys with composition Fe-(10-13)Cr-(5-7)Al in wt.% at temperatures of 214, 357, and 557°C to a dose of ~1.8 dpa (displacements per atom). Dislocation loop sizes and dispersion characteristics did not show strong compositional dependence. However, there were noticeable effects of composition on the dispersion of Cr-rich α'-precipitates, particularly a decrease in number density with increasing Al content or a decrease in Cr content. The present results confirm previous studies indicating the Cr concentration in these precipitates is lower than that expected in binary FeCr alloys and that Al can act as a destabilizing alloying element for the deleterious α'-phase.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Solid-state dewetting instability in thermally-stable nanocrystalline binary alloys

Practical applications of nanocrystalline metallic thin films are often limited by instabilities. Furthermore, in addition to grain growth, the thin film itself can become unstable and collapse into islands through solid-state dewetting. Additionally, selective alloying can improve nanocrystalline stability, but the impact of this approach on dewetting is not clear. In this study, two alloys that exhibit nanocrystalline thermal stability as ball milled powders are evaluated as thin films. While both alloys demonstrated dewetting behavior following annealing, the severity decreased in more dilute compositions. Ultimately, a balance may be struck between nanocrystalline stability and thin film structural stability by tuning dopant concentration.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Investigations of irradiation effects in crystalline and amorphous SiC

The results of irradiation on 3C-silicon carbide (SiC) and amorphous SiC (a-SiC) are explored using both in situ transmission electron microscopy (TEM) and complementary molecular dynamics (MD) simulations. The single ion strikes identified in the in situ TEM irradiation experiments, utilizing a 1.7 MeV Au 3+ ion beam with nanosecond resolution, are contrasted to MD simulation results of the defect cascades produced by 10–100 keV Si primary knock-on atoms (PKAs). The MD simulations also investigated defect structures that could possibly be responsible for the observed strain fields produced by single ion strikes in the TEM ion beam irradiation experiments. Both MD simulations and in situ TEM experiments show evidence of radiation damage in 3C-SiC but none in a-SiC. Selected area electron diffraction patterns, due to the results of MD simulations and in situ TEM irradiation experiments, show no evidence of structural changes in either 3C-SiC or a-SiC.

3C-SiC↗
DOE OSTI2019

Influence of welding and neutron irradiation on dislocation loop formation and α' precipitation in a FeCrAl alloy

An advanced accident-tolerant FeCrAl alloy, C35 M (Fe–13Cr–10Al–1Mo, at %), and its laser-fusion weldments were studied after neutron irradiation up to 1.8 dpa at 357 °C to evaluate the radiation damage and α'-precipitation kinetics. Results show that the densities of α' in the fusion zone (FZ) and the heat-affected zone (HAZ) were only a third of that in the base material, i.e., 2.3 × 10 24 m -3 . Here, density of dislocation loops in the FZ and HAZ were less than half of that in base material. The size of α' precipitates and the dislocation loops were found to be smaller in base material, compared to FZ and HAZ. The variances in microstructural response within the different regions were attributed to initial as-welded states of each region, indicating the impact of the pre-irradiation condition on radiation response in FeCrAl alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Role of refractory inclusions in the radiation-induced microstructure of APMT

Kanthal APMT is a promising FeCrAl-based alloy for accident-tolerant fuel cladding because of its excellent high-temperature oxidation resistance. In this study, powder metallurgy Kanthal APMT alloy, neutron irradiated to 1.8 dpa at nominally 382 °C, was characterized. On-zone STEM imaging revealed that radiation-induced dislocation loops with Burgers vectors of a/2 <111 > or and black dots tended to agglomerate in the vicinity of refractory inclusions. The densities and sizes of these loops decreased with distance from the inclusion-matrix interfaces. In addition, high-resolution energy-dispersive X-ray spectroscopy mapping was used to determine the inclusions to be either yttrium- or silicon-rich, as well as to detect the radial distribution of radiation-enhanced α' phase near these inclusions. A high density of randomly distributed Cr-rich α' phase was found, regardless of the presence of inclusions. Here, the results from this study provide insights into how microstructural features can locally tailor the radiation-induced defects in FeCrAl-based alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Precipitation of α' in neutron irradiated commercial FeCrAl alloys

In this paper, Alkrothal 720 and Kanthal APMT™, two commercial FeCrAl alloys, were neutron irradiated up to damage doses of 7.0 displacements per atom (dpa) in the temperature range of 320 to 382 °C to characterize the α' precipitation in these alloys using small-angle neutron scattering. Both alloys exhibited α' precipitation. Kanthal APMT™ exhibited higher number densities and volume fraction, a result attributed to its higher Cr content compared with Alkrothal 720. Finally, trends observed as a function of damage dose (dpa) are consistent with literature trends for both FeCr and FeCrAl alloys

36 MATERIALS SCIENCE↗
DOE OSTI2017

Dislocation loop formation in model FeCrAl alloys after neutron irradiation below 1 dpa

FeCrAl alloys with varying compositions and microstructures are under consideration for accident-tolerant fuel cladding, but limited details exist on dislocation loop formation and growth for this class of alloys under neutron irradiation. Four model FeCrAl alloys with chromium contents ranging from 10.01 to 17.51 wt % and alunimum contents of 4.78 to 2.93 wt % were neutron irradiated to doses of 0.3–0.8 displacements per atom (dpa) at temperatures of 335–355°C. On-zone STEM imaging revealed a mixed population of black dots and larger dislocation loops with either a/2< 111 > or a< 100 > Burgers vectors. Weak composition dependencies were observed and varied depending on whether the defect size, number density, or ratio of defect types was of interest. Here, the results were found to mirror those of previous studies on FeCrAl and FeCr alloys irradiated under similar conditions, although distinct differences exist.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Dislocation loop evolution during in-situ ion irradiation of model FeCrAl alloys

Model FeCrAl alloys of Fe-10%Cr-5%Al, Fe-12%Cr-4.5%Al, Fe-15%Cr-4%Al, and Fe-18%Cr-3%Al (in wt %) were irradiated with 1 MeV Kr++ ions in-situ with transmission electron microscopy to a dose of 2.5 displacements per atom (dpa) at 320 °C. In all cases, the microstructural damage consisted of dislocation loops with ½< 111 > and <100 > Burgers vectors. The proportion of ½< 111 > dislocation loops varied from ~50% in the Fe-10%Cr-5%Al model alloy and the Fe-18Cr%-3%Al model alloy to a peak of ~80% in the model Fe-15%Cr-4.5%Al alloy. The dislocation loop volume density increased with dose for all alloys and showed signs of approaching an upper limit. The total loop populations at 2.5 dpa had a slight (and possibly insignificant) decline as the chromium content was increased from 10 to 15 wt %, but the Fe-18%Cr-3%Al alloy had a dislocation loop population ~50% smaller than the other model alloys. As a result, the largest dislocation loops in each alloy had image sizes of close to 20 nm in the micrographs, and the median diameters for all alloys ranged from 6 to 8 nm. Nature analysis by the inside-outside method indicated most dislocation loops were interstitial type.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Mechanical properties of neutron-irradiated model and commercial FeCrAl alloys

The development and understanding of the mechanical properties of neutron-irradiated FeCrAl alloys is increasingly a critical need as these alloys continue to become more mature for nuclear reactor applications. This study focuses on the mechanical properties of model FeCrAl alloys and of a commercial FeCrAl alloy neutron-irradiated to up to 13.8 displacements per atom (dpa) at irradiation temperatures between 320 and 382 °C. Tensile tests were completed at room temperature and at 320 °C, and a subset of fractured tensile specimens was examined by scanning electron microscopy. Results showed typical radiation hardening and embrittlement indicative of high chromium ferritic alloys with strong chromium composition dependencies at lower doses. At and above 7.0 dpa, the mechanical properties saturated for both the commercial and model FeCrAl alloys, although brittle cleavage fracture was observed at the highest dose in the model FeCrAl alloy with the highest chromium content (18 wt %). Finally, the results suggest the composition and microstructure of FeCrAl alloys plays a critical role in the mechanical response of FeCrAl alloys irradiated near temperatures relevant to light water reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2017

A combined APT and SANS investigation of α' phase precipitation in neutron-irradiated model FeCrAl alloys

Here, FeCrAl alloys are currently under consideration for accident-tolerant fuel cladding applications in light water reactors owing to their superior high-temperature oxidation and corrosion resistance compared to the Zr-based alloys currently employed. However, their performance could be limited by precipitation of a Cr-rich α' phase that tends to embrittle high-Cr ferritic Fe-based alloys. In this study, four FeCrAl model alloys with 10–18 at.% Cr and 5.8–9.3 at.% Al were neutron-irradiated to nominal damage doses up to 7.0 displacements per atom at a target temperature of 320 °C. Small angle neutron scattering techniques were coupled with atom probe tomography to assess the composition and morphology of the resulting α' precipitates. It was demonstrated that Al additions partially destabilize the α' phase, generally resulting in precipitates with lower Cr contents when compared with binary Fe-Cr systems. The precipitate morphology evolution with dose exhibited a transient coarsening regime akin to previously observed behavior in aged Fe-Cr alloys. Similar behavior to predictions of the LSW/UOKV models suggests that α' precipitation in irradiated FeCrAl is a diffusion-limited process with coarsening mechanisms similar to those in thermally aged high-Cr ferritic alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Correlative Microscopy of Neutron-Irradiated Materials

A nuclear reactor core is a highly demanding environment that presents several unique challenges for materials performance. Materials in modern light water reactor (LWR) cores must survive several decades in high-temperature (300-350°C) aqueous corrosion conditions while being subject to large amounts of high-energy neutron irradiation. Next-generation reactor designs seek to use more corrosive coolants (e.g., molten salts) and even greater temperatures and neutron doses. The high amounts of disorder and unique crystallographic defects and microchemical segregation effects induced by radiation inevitably lead to property degradation of materials. Thus, maintaining structural integrity and safety margins over the course of the reactor's service life thus necessitates the ability to understand and predict these degradation phenomena in order to develop new, radiation-tolerant materials that can maintain the required performance in these extreme conditions.

36 MATERIALS SCIENCE↗